Efficiency analyses of the CANDU spent fuel repository using modified disposal canisters for a deep geological disposal system design

Deep geological disposal concept is considered to be the most preferable for isolating high-level radioactive waste (HLW), including nuclear spent fuels, from the biosphere in a safe manner. The purpose of deep geological disposal of HLW is to isolate radioactive waste and to inhibit its release of for a long time, so that its toxicity does not affect the human beings and the biosphere. One of the most important requirements of HLW repository design for a deep geological disposal system is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. In this study, a reference disposal concept for spent nuclear fuels in Korea has been reviewed, and based on this concept, efficient alternative concepts that consider modified CANDU spent fuels disposal canister, were developed. To meet the thermal requirement of the disposal system, the spacing of the disposal tunnels and that of the disposal pits for each alternative concept, were drawn following heat transfer analyses. From the result of the thermal analyses, the disposal efficiency of the alternative concepts was reviewed and the most effective concept suggested. The results of these analyses can be used for a deep geological repository design and detailed analyses, based on exact site characteristics data, will reduce the uncertainty of the results.     

Availability of a probabilistic cost estimation for the price effect of Cu powder and bentonite on an HLW disposal cost in Korea

This study aims to demonstrate the availability of a probabilistic cost estimation related to the price effects of Cu powder and bentonite. From a sensitivity analysis of those materials’ prices on the overall disposal costs, it was found that Cu powder was a more dominant cost driver than that of bentonite among the material costs to dispose of 52,000 tU of spent fuels by the deterministic cost estimation method even though the used volume of Cu powder will be smaller than that of bentonite, whereas those conclusions can be changed by a probabilistic cost estimation method. Namely, its conclusion depends on a decision maker's personal opinion because of the resultant uncertainties. The disposal cost includes too many uncertainties due to the long construction and operational durations of a repository. Therefore a probabilistic cost estimation can be useful to provide the information related to an uncertainty.     

Preparation,characterisation and dissolution of a CeO2 analogue for UO2 nuclear fuel

The behaviour of spent nuclear fuel under geological conditions is a major issue underpinning the safety case for final disposal. This work describes the preparation and characterisation of a non-radioactive UO₂ fuel analogue, CeO₂, to be used to investigate nuclear fuel dissolution under realistic repository conditions as part of a developing EU research programme. The densification behaviour of several cerium dioxide powders, derived from cerium oxalate, were investigated to aid the selection of a suitable powder for fabrication of fuel analogues for powder dissolution tests. CeO₂ powders prepared by calcination of cerium oxalate at 800 °C and sintering at 1700 °C gave samples with similar microstructure to UO₂ fuel and SIMFUEL. The suitability of the optimised synthesis route for dissolution was tested in a dissolution experiment conducted at 90 °C in 0.01 M HNO₃.     

Suction and water uptake in unsaturated compacted bentonite

Suction is an important process dominating water movement in unsaturated compacted bentonite. Suction measurements were carried out to investigate the characteristics of suction in unsaturated compacted bentonite for the buffer of a HLW repository. The suction values decreased with increasing the water content at a given dry density, and it revealed higher value at the higher dry density of compacted bentonite. The suction variation with an increase of temperature was negligible for the water content up to 17%. For the dry density of 1600 kg/m³, the suction value in the vertical measurement was a little lower than that in the horizontal one due to the microstructural anisotropy of compacted bentonite. The modeling of water uptake in unsaturated compacted bentonite is allowed to predict the re-saturation for the buffer of a HLW repository. To this end, water uptake tests were conducted to compare their results with those calculated using the coupled hydro-mechanical model of a computer code ABAQUS. Both results were in reasonably good agreement with each other, which suggests that the model can be used to predict the re-saturation of the bentonitic buffer in a HLW repository.     

Advanced fuel cycle scenarios for High Temperature Gas Reactors

This study evaluates nuclear fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. Three fuel cycle scenarios, the Light Water Reactor (LWR)–Advanced Recycling Reactor (ARR) recycle, the LWR–High Temperature Gas Reactor (HTGR)–ARR recycle, and the HTGR partial recycling fuel cycle, are assessed for their mass flow and electricity generation costs and the results are compared to those of the LWR once-through fuel cycle. The spent fuels are recycled in both the Consolidated Fuel Treatment Center and the Actinide Management Island, which are capable of reprocessing spent fuels by Uranium Extraction and Pyrochemical processes, respectively. The mass flow calculations show that the Transuranics (TRU) which have a long-term radiation effect can be completely burned in the recycling fuel cycles, resulting in 350, 450 and 6 times reduction of TRU inventory for the LWR–ARR, LWR–HTGR–ARR and HTGR partial recycling fuel cycles, respectively, when compared to the once-through fuel cycle. The electricity generation costs of these fuel cycle scenarios were estimated to be 39.1, 34.9 and 25.7 USD/MW h(e), which are comparable to or smaller than that of the once-through fuel cycle. Although the candidate fuel cycles adopt reprocessing options which raise fuel cycle cost, increase in uranium cost and the advanced design of the HTGR can further reduce the advanced fuel cycle costs of the HTGR.     

Long-term Behavior Science: The cornerstone approach for reliably assessing the long-term performance of nuclear waste

High level waste (HLW) management requires being able to demonstrate the safety over geological timescales, typically 1 My. This can be made possible by using a rigorous, complex and iterative scientific approach called Long-term Behavior Science. The methodology relies on experiments and modelling. A large multi-scale approach is required and involves a mechanistic understanding of the key phenomena controlling the source term (i.e. the flux of radionuclides released from the waste as a function of time), as well as parametric studies, integrated and in situ tests. As a result, it is eventually possible to develop an operational model based on clever simplifications of a very complex reality, ensuring that predictions will always remain conservative despite conceptual and numerical uncertainties. Finally, predictive models must be validated based on the study of natural or archaeological analogues. In this paper, we show how this methodology can be applied by selecting examples of spent nuclear fuel and HLW glass.     

Burnup study for Pakistan Research Reactor-1 utilizing high density low enriched uranium fuel

Burnup study for Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type MTR utilizing high density low enriched uranium fuel, was performed by using Fuel Cycle Analysis Program (FCAP). Existing equilibrium core of PARR-1, which is relatively economical but provides less neutron fluxes per unit power than the first equilibrium core, was formed by adding five more fuel elements in the first equilibrium core. This study shows that if the fuel loading is increased in the first equilibrium core of PARR-1 by replacing the fuel of density 3.28 gU/cm³ by the fuel of density 4.00 gU/cm³ then the new equilibrium core can provide 10% higher neutron fluxes at the irradiation sites and will also require 1.5 kg less fuel than that required for existing equilibrium core for one-year full power operation at 10 MW. The new core provides neutron fluxes at 13% lower cost and if the size of this core is further reduced by three fuel elements then this core can provide 20% higher thermal neutron flux at the central flux trap at 9% lower cost. A possible use of U-Mo (5 w/o Mo) fuel of density 8.5 gU/cm³ in PARR-1 with an increase in existing water channel width from 2.1 to 2.45 mm (Ann. Nucl. Energy 32(1), 29–62) would provide up to 41% more thermal neutron flux at the central flux trap at 13% lower cost than the existing equilibrium core. The power peaking factors in these cores are similar to the power peaking factors of the existing equilibrium core and these cores are likely to operate within the safety constraints as defined for the existing equilibrium core of PARR-1.     

Development and application of high volume remote handling systems in support of JET and ITER

The Joint European Torus (JET) Remote Handling System has evolved from a small scale maintenance capability to one of high efficiency large volume installations. The Enhanced Performance 2 shutdown 2010–2011 for example, required the replacement of many thousands of components ranging from about 100 g to 130 kg in weight. The scale of this type of operation and the necessity to maximise operational availability intensified the demands for high productivity whilst maintaining the necessary high standards for precision, reliability, cleanliness, and operational security.This paper discusses the developments in design, control, maintenance, preparation and operation of the current state of the art remote handling facilities at JET. It explores how the experience of over 20,000 h of operations has developed the applied methodology and how this could be appropriate to ITER and other facilities requiring complex remote maintenance, where extensive, high productivity remote handling operations will be essential. It also discusses the advances that have been made in management and presentation of operational data within the command, control and human machine interfaces (HMI) systems, along with the supporting operational databases.     

The use of NRA to study thermal diffusion of helium in (U, Pu)O2

A lot of work has been already done on helium atomic diffusion in UO₂ samples, but information is still lacking about the fate of helium in high level damaged UOX and MOX matrices and more precisely their intrinsic evolutions under alpha self irradiation in disposal/storage conditions.The present study deals with helium atomic diffusion in actinide doped samples versus damage level. The presently used samples allow a disposal simulation of about 100 years of a UOX spent fuel with a 60 MW d kg^?1 burnup or a storage simulation of a MOX spent fuel with a 47.5 MW d kg^?1 burnup.For the first time, nuclear reaction analysis of radioactive samples has been performed in order to obtain diffusion coefficients of helium in (U, Pu)O₂. Samples were implanted with ³He⁺ and then annealed at temperatures ranging from 1123 K to 1273 K. The evolution of the ³He depth profiles was studied by the mean of the non-resonant reaction: ³He(d, p)⁴He. Using the SIMNRA software and the second Fick’s law, thermal diffusion coefficients have been measured and compared to the ³He thermal diffusion coefficients in UO₂ found in the literature.     

Conceptual design of compact supercritical water-cooled fast reactor with thermal hydraulic coupling

Fuel rod design for high power density supercritical water-cooled fast reactor was conducted with mixed-oxide (MOX) fuel and stainless steel (SUS304) cladding under the limiting cladding surface temperature of 650 °C. Fuel and cladding integrities, and flow-induced vibration were taken into account as design criteria. Designed fuel rod has the diameter of 7.6 mm and is arranged in the fuel assembly with pitch-to-diameter ratio of 1.14. New core arrangement for negative void reactivity is proposed by three-dimensional tri-z core calculation fully coupled with thermal hydraulic calculation, where ZrH layer concept is used for negative void reactivity. The core has high power density of 156 W/cm³ and its equivalent diameter is only 2.7 m for 1000 MWe class reactor core. High average core outlet temperature of 500 °C is achieved by introducing radial fuel enrichment zoning and downward flow in seed assembly. Small pressure vessel size and simplified direct steam cycle with higher thermal efficiency give an economical potential in aspect of capital and operating cost.     

Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.     

TRU self-recycling in a high temperature modular helium reactor

The Deep Burn Project is developing high burnup fuel based on Ceramically Coated (TRISO) particles, for use in the management of spent fuel Transuranics. This paper evaluates the TRU deep-burn in a High Temperature Reactor (HTR) that recycles its own transuranic production. The DB-HTR is loaded with standard LEU fresh fuel and the self-generated TRUs are recycled into the same core (after reprocessing of the original spent fuel). This mode of operation is called self-recycling (SR-HTR). The final spent fuel of the SR-HTR can be disposed of in a final repository, or recycled again.In this study, a single recycling of the self-generated TRUs is considered. The UO₂ fuel kernel is 12% uranium enrichment and the diameter of the kernel is 500 μm. TRISO packing fraction of UO₂ fuel compact is 26%. In the SR-HTR fuel cycle, it is assumed that the spent UO₂ fuel is reprocessed with conventional technology and the recovered TRUs are fabricated into Deep Burn TRISO fuel. The diameter of 200 μm is used for the TRU fuel kernel. A typical coating thickness is used. The core performance is evaluated for an equilibrium cycle, which is obtained by cycle-wise depletion calculations. From the analysis results, the equilibrium cycle lengths of Case 1 (5-ring fuel block SR-HTR) and Case 2 (4-ring fuel block SR-HTR) are 487 and 450 EFPDs (effective full power days), respectively. And the UO₂ fuel discharge burnups of Case 1 and Case 2 are 10.3% and 10.1%, respectively. Also, the TRU discharge burnups of Case 1 and Case 2 are 64.7% and 63.5%, respectively, which is considered extremely high. The fissile (Pu-239 and Pu-241) content of the self-generated TRU vector is about 52%. The deep-burning of TRU in SR-HTR is partly due to the efficient conversion of Pu-240 to Pu-241, which is boosted by the uranium fuel in SR-HTR. It is also observed that the power distribution is quite flat within the uranium fuel zone. The lower power density in TRU fuel is because the TRU burnup is very high. Also, it is found that transmutation of Pu-239 is near complete in SR-HTR and that of Pu-241 is extremely high in all cases. The decay heat of the SR-HTR core is very similar to the UO₂-only core. However, accumulation of the minor actinides is not avoidable in the SR-HTR core. The extreme high burnup of the Deep Burn fuel greatly reduces the amount of heat producing isotopes that could be problematic in spent fuel repositories (like Pu-238).     

Spalling effect on gamma-ray shielding performance in MACSTOR-400 (Korean CANDU spent fuel storage module)

The saturation of South Korea's at-reactor (AR) spent fuel storage pools has created a need for additional spent fuel storage capacity. The Korea Hydro & Nuclear Power Company is planning to construct a MACSTOR-400 facility composed of reinforced concrete.In concrete structures, cracks occur due to thermal stress, hydration heat, weather, load and other reasons, and shielding performance changes according to the event by crack. However, there are no researches providing the allowable crack for shielding. This paper presents estimating the effect of concrete spalling on gamma-ray shielding performance in MACSTOR-400 based on an FEM simulation. When the cover thickness is under 10 cm, it is estimated that spalling will be formed. Generally, because cover thickness is under 10 cm in the concrete structure, it is estimated that spalling will be formed in MACSTOR-400. If the thickness of wall in MACSTOR-400 is over 100 cm, it is estimated that surface dose rate will be maintained within the radiation safety standards under any conditions related to crack effect. The results of this study may be used for developing standards for radiological safety in shielding structures, especially in relation to the shielding margin.     

A model for coolant void reactivity evaluation in assemblies of CANDU cells

Coolant void reactivity is a very important safety parameter in CANDU reactor analysis. Here we evaluate the coolant void reactivity in a 2 × 2 heterogeneous assembly of CANDU cells using the code DRAGON. Since the current version of DRAGON can only treat the coolant void reactivity for a single CANDU cell, an approximate model for the geometry must be considered to perform assembly calculations in a 2 × 2 pattern. The model we propose consists of replacing the annular fuel pins by equivalent square fuel pins. The equivalence between annular fuel pins and square fuel pins is brought about by homogenizing the fuel plus its sheath and subsequently conserving the reaction rates between the two geometries using a SPH equivalence procedure. The approximate CANDU cells constructed using square pins were used to perform the transport calculations in 2 × 2 assembly patterns. In addition, the model was used to evaluate coolant void reactivity in 2 × 2 checkerboard voiding patterns. These calculations reflect more accurately the actual voiding situation being studied. This helps in assessing the effects due to the coupling of neutrons born in one cell to those born in the neighbouring cells.     

Rapid aqueous release of fission products from high burn-up LWR fuel: Experimental results and correlations with fission gas release

Studies of the rapid aqueous release of fission products from UO₂ and MOX fuel are of interest for the assessment of the safety of geological disposal of spent fuel, because of the associated potential contribution to dose in radiological safety assessment. Studies have shown that correlations between fission gas release (FGR) and the fraction rapidly leached of various long-lived fission products can provide a useful method to obtain some of this information. Previously, these studies have been limited largely to fuel with burn-up values below 50 MWd/kg U. Collaborative studies involving SKB, Studsvik, Nagra and PSI have provided new data on short-term release of ¹³⁷Cs and ¹²⁹I for a number of fuels irradiated to burn-ups of 50–75 MWd/kgU. In addition a method for analysis of leaching solutions for ⁷⁹Se was developed. The results of the studies show that the fractional release of ¹³⁷Cs is usually much lower than the FGR covering the entire range of burn-ups studied. Fractional ¹²⁹I releases are somewhat larger, but only in cases in which the fuel was forcibly extracted from the cladding. Despite the expected high degree of segregation of fission gas (and by association ¹³⁷Cs and ¹²⁹I) in the high burn-up rim, no evidence was found for a significant contribution to release from the rim region. The method for ⁷⁹Se analysis developed did not permit its detection. Nonetheless, based on the detection limit, the results suggest that ⁷⁹Se is not preferentially leached from spent fuel.     

Prospects of using different clad materials in a material test research reactor – Part 3 – The dynamic behavior

The effects of using different clad materials on the dynamics of a material test research reactor were studied. For this purpose, the aluminum clad of an MTR was replaced separately with stainless steel-316 and zircaloy-4. Simulations were carried out to determine the reactor performance under reactivity insertion and loss-of-flow transients. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that during the fast reactivity insertion transient, Al cladded fuel attained the maximum reactor power of 59.34 MW, while stainless steel-316 cladded attained 48.74 MW and zircaloy-4 cladded attained maximum power of 55.87 MW. During the slow reactivity insertion transient, Al cladded fuel attained the maximum reactor power of 12.38 MW, while stainless steel-316 cladded attained 12.23 MW and zircaloy-4 cladded attained maximum power of 12.34 MW. During the loss-of-flow transients, the reactor power of the stainless steel-316 cladded fuel remained slightly lower than the other two. The fuel temperature of stainless steel-316 and zircaloy-4 cladded fuels remained higher due to poor fuel–clad gap conductance.     

Corrosion study of a highly durable electrolyzer based on cold crucible technique for pyrochemical reprocessing of spent nuclear oxide fuel

The application of the cold crucible technique to a pyrochemical electrolyzer used in the oxide-electrowinning method, which is a method for the pyrochemical reprocessing of spent nuclear oxide fuel, is proposed as a means for improving corrosion resistance. The electrolyzer suffers from a severe corrosion environment consisting of molten salt and corrosive gas. In this study, corrosion tests for several metals in molten 2CsCl–NaCl at 923 K with purging chlorine gas were conducted under controlled material temperature conditions. The results revealed that the corrosion rates of several materials were significantly decreased by the material cooling effect. In particular, Hastelloy C-22 showed excellent corrosion resistance with a corrosion rate of just under 0.01 mm/y in both molten salt and vapor phases by controlling the material surface at 473 K. Finally, an engineering-scale crucible composed of Hastelloy C-22 was manufactured to demonstrate the basic function of the cold crucible. The cold crucible induction melting system with the new concept Hastelloy crucible showed good compatibility with respect to its heating and cooling performances.     

Neutronic model of a mirror based fusion–fission hybrid for the incineration of the transuranic elements from spent nuclear fuel and energy amplification

The Georgia Institute of Technology has developed several design concepts of tokamak based fusion–fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a “fusion transmutation of waste reactor”. It was designed for a criticality k_eff ≤ 0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GW_th requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBR_cycle = 1.9 and a first wall lifetime of 12–16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead–bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality k_eff ≤ 0.97 and produces less fission power of 1.5 GW_th. Its main performance parameters turn out substantially better.     

Assessment of fuel conversion from HEU to LEU in the Syrian MNSR reactor using the MCNP code

Assessment of fuel conversion from high enriched uranium (HEU) to low enriched uranium (LEU) fuel in the Syrian MNSR reactor was conducted in this paper. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed to assess the possibility of fuel conversion from 89.87% HEU fuel (UAl₄–Al) to 19.75% LEU fuel (UO₂). The first model showed that 347 fuel rods with HEU fuel were required to obtain a reactor core with 5.17 mk unadjusted excess reactivity. The second model showed that only 200 LEU fuel rods distributed in the reactor core like the David star figure were required to obtain a reactor core with 4.85 mk unadjusted excess reactivity. The control rod worth using the LEU fuel was enhanced. Finally, the third model showed that distribution of 200 LEU fuel rods isotropically in the 10 circles of the reactor core failed to convert the fuel since the calculated core unadjusted excess reactivity for this model was 10.45 mk. This value was far beyond the reactor operation limits and highly exceeded the current MNSR core unadjusted excess reactivity (5.17 mk).     

Analyses of disposal efficiency based on nuclear spent fuel cooling time and disposal tunnel/pit spacing for the design of a geological repository

The purpose of deep geological disposal of high-level radioactive waste (HLW) including nuclear spent fuels is to isolate and to inhibit the release of radioactive material for a long time so that its toxicity does not affect the biosphere. The main requirement for the HLW repository design is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. The cooling time of the spent fuels discharged from nuclear power plants is the key consideration factor for the efficiency and economic feasibility of such a repository. We analyze the spacing of the disposal tunnels and pits, the disposal area and the uranium density for the deep geological repository layout to satisfy the thermal requirement of the disposal system. To do this, thermal stability analyses of a disposal system have been performed using varying spent fuel cooling times and spacing of the disposal tunnels and pits. The results show that the time to reach the maximum temperature within the design limit of the temperature in the disposal site is likely to be shortened as the cooling time of the spent fuel becomes shorter. Also it seems that controlling the disposal pit spacing is considered more advantageous than controlling the disposal tunnel spacing to meet the allowable thermal criteria in the repository from thermal and economical points of view. The results of these analyses can be used for a deep geological repository design and detailed analyses with exact site characteristics data will reduce the uncertainty of the results.